1. Field of the Invention
This invention concerns a model steam generator having an improved feedwater system which allows the operator to test the efficacy of selected anti-corrosive additives and chemical cleaning agents in the boiler feedwater.
2. Description of the Prior Art
Model steam generators for monitoring the amount of corrosion degradation occurring within the heat exchange tubes of a nuclear steam generator are known in the prior art. Generally speaking, such model generators operate by subjecting an array of sample heat exchange tubes to the same heat, pressure and chemical conditions which surround the heat exchange tubes in nuclear steam generators. If these conditions are accurately simulated, the amount of corrosion which occurs in the sample tubes of the model steam generator will provide an accurate indication of the tube corrosion present in the nuclear steam generator being monitored. Such model steam generators are a particularly useful form of corrosion monitor, because they obviate the need for shutting down the nuclear plant and sending technicians into the radioactive interiors of the generators. However, such model steam generators are useful only insofar as they are capable of accurately simulating the heat, pressure and chemical conditions which exist inside the nuclear plant. Any material departures from these conditions will adversely affect the accuracy of the model steam generator.
In order to understand the difficulties in building a practical model steam generator which provides an accurate monitor for heat exchange tube corrosion, one must first understand how nuclear steam generators are generally constructed, and what chemical and hydraulic conditions are responsible for tube corrosion.
Nuclear steam generators are comprised of three principal parts, including a secondary side and a tubesheet, as well as a primary side which circulates water heated from a nuclear reactor. The secondary side of the generator includes a plurality of U-shaped tubes, as well as an inlet for admitting a flow of feedwater. The inlet and outlet ends of the U-shaped tubes within the secondary side are mounted in the tubesheet which hydraulically separates the primary side of the generator from the secondary side. The primary side in turn includes a divider sheet which hydraulically isolates the inlet ends of the U-shaped tubes from the outlet ends. Hot water flowing from the nuclear reactor is admitted into the section of the primary side containing all of the inlet ends of the U-shaped tubes. This hot water flows through these inlets, up through the tubesheet, and circulates around the U-shaped tubes which extend within the secondary side of the steam generator. The heated water transfers its heat through the walls of the U-shaped tubes to the feedwater flowing through the secondary side of the generator, thereby converting the feedwater to steam. After the nuclear-heated water circulates through the U-shaped tubes, it flows back through the tubesheet, through the outlets of the U-shaped tubes, and into the outlet section of the primary side, where it is recirculated back to the nuclear reactor. The inlet ends of the U-shaped tubes are known as the "hot legs", and the outlet ends of these tubes are known as the "cold legs".
The heat exchange tubes of such nuclear steam generators can suffer a number of different types of corrosion degradation, including denting, stress corrosion cracking, intragranullar attack, and pitting, In situ examination of the tubes within these generators has revealed that most of this corrosion degradation occurs in what are known as the crevice regions of the generator. Such crevice regions include the annular space between the heat exchange tubes and the tubesheet, as well as the annular clearance between these tubes and the various support plates in the secondary side which are used to uniformly space and align these tubes. Corrosive sludge tends to collect within these crevices from the effects of gravity. Moreover, the relatively poor hydraulic circulation of the water in these regions tends to maintain the sludge in these crevices, and to create localized "hot spots" in the tubes adjacent the sludge. The heat radiating from these "hot spots" acts as a powerful catalyst in causing the exterior surface of the heat exchange tubes to chemically combine with the corrosive chemicals in the sludge. While most nuclear steam generators include blow-down systems for periodically sweeping the sludge out of the generator vessel, the sludges in the crevice regions are not easily swept away by the hydraulic currents induced by such systems. Despite the fact that the heat exchange tubes of such nuclear generators are typically formed from corrosion-resistant Inconel stainless steel, the combination of the localized regions of heat and corrosive sludges can utilimately cause the heat exchange tubes to crack, the leak radioactive water from the primary side into the secondary side of the generator. However, this need not occur if the heat exchange tubes are provided with internally reinforcing sleeves before the corrosion causes cracks in the tube walls.
Model steam generators were developed in order to accurately monitor the amount of corrosion degradation occurring in the heat exchange tubes of a particular nuclear steam generator, in order that these tubes might be sleeved before any of the tube walls crack. Such model steam generators have been found to be a particularly accurate way of ascertaining the amount of corrosion degradation occurring in the heat exchange tubes of a nuclear steam generator, because the particular amount of corrosion which the feedwater chemistry and thermohydraulics of the particular generator will induce in a particular set of tubes is virtually impossible to predict by purely theoretical models.
However, such prior art model steam generators are not without significant shortcomings. For example, feedwater inlet systems of prior art model generators make it difficult, if not impossible, to easily test the effects of selected anti-corrosive additives and cleaning agents in the feedwater. The inability to easily test the efficacy of such additives and cleaning agents is a serious deficiency, since the use of the proper type and amount of such addivites could save the operators of nuclear steam generators many thousands of dollars in maintenance costs. Additionally, such prior art systems often allow the flow of feedwater en route to the model steam generator from the nuclear steam generator to slow down to the point where particulate solids entrained in this feedwater settle out in the inlet conduits before this feedwater is introduced into the secondary side of the model steam generator. Stated differently, such prior art model steam generators have no means for insuring that the flow through the feedwater inlet conduit will be sufficiently turbulent over the entire generator operating range so as to insure that particulate solids in the feedwater flowing into the nuclear steam generator will also flow into the model steam generator. If such particulate solids are allowed to settle out, the ability of the model steam generator to accurately simulate the amount of corrosive sludge accumulation which occurs within the nuclear steam generator will be substantially impaired, thus materially affecting the accuracy of the monitoring ability of the model steam generator.
Clearly, a need exists for a feedwater inlet system which would allow anti-corrosive additives and cleaning agents to be easily and conveniently mixed within the feedwater flowing into the model steam generator, in order that the efficacy of such chemicals may be accurately determined. Such a system should also be capable of maintaining a turbulent flow within the entire length of the inlet conduit in order that particulate solids entrained in the feedwater will not have an opportunity to settle before this feedwater is introduced into the secondary side of the generator. Such ability would materially contribute to the capacity of the model steam generator to accurately simulate the sludge-building characteristics of the feedwater used in the nuclear steam generator.